GIOVANNI LARANJO DE STEFANI

GIOVANNI LARANJO DE STEFANI

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The AP-Th 1000 an advanced concept to use MOX of thorium in a closed fuel cycle
2019 - STEFANI, GIOVANNI L. de; MAIORINO, JOSE R.
This work presents a study for the firsts 4 cycles of recharge of the reactor AP-Th1000, a version of the reactor AP1000 using mixed uranium and thorium oxides as fuel, which the feasibility studies had been already demonstrated in previous study for a first cycle. The AP-Th 1000 study is a proposal to start the thorium fuel cycle using the most common reactor technology in the nuclear industry, the Pressurized Water Reactors (PWR). A preliminary closed cycle study is carried out for the first 4 cycles where the production of 233U are evaluated. In cycles 2, 3 and 4, new assemblies with a fuel of the remaining uranium from the previous cycle are used instead of assemblies removed from the core, thus being a mixture of different uranium’s (232U, 233U, 234U, 235U, 236U and 238U) , where the additional fissile material inserted into the fuel to ensure the 18-month operation of the reactor comes from uranium oxides enriched at 20 w / o.. The results demonstrate the viability of the proposal and again using closed fuel cycle.
The AP-Th 1000
2021 - STEFANI, GIOVANNI L. de; MAIORINO, JOSE R.; MOREIRA, JOAO M. de L.
This work presents the feasibility studies to convert the UO2 core of the Westinghouse AP1000 reactor to a U/Th core aiming at U/Th fuel recycling. The focus of the work is to establish a first core which allows normal operation of the AP1000 reactor and investigate a possible route for generating the 233U for U/Th fuel recycling. The converted core named AP-Th1000 is divided in three homogenous zones with different UO2/ThO2 mass proportions. The reprocessing procedure envisioned is to separate fission products and Pu isotopes, retain Uranium, use this fuel material in subsequent fuel cycles and complement the required fissile material with U with enrichment below 20%. The goal was to gradually reduce the mass proportion of mined Uranium fuel and eventually attain a Th-233U core with similar operation characteristics of current AP1000 core. We perform a detailed three-dimensional full core analysis with the SERPENT code examining core reactivity, power density distribution, and also a preliminary closed cycle study for the first 4 cycles where the production of 233U are evaluated. The goal of converting the AP1000 reactor core to a U/ThO2 fuel cycle was partially accomplished. While the first cycle was thoroughly examined and met all requirements we were not able to find a route to migrate it to a prevalent Th cycle. Basically, two of the set of criteria adopted in the study proved to be too restrictive to attain this goal with homogenous assembly, namely U enrichment below 20% and not recycling Pu. The results indicate that removing these two criteria the conversion factor in the ensuing fuel cycles can be increased and possibly attain a Th cycle without compromising the economics of power generation. The design changes were the elimination of IFBA burnable absorbers and replacement of gray control bundles by black control bundles.
STC-MOX-Th
2020 - SANTOS, THIAGO A. dos; STEFANI, GIOVANNI L. de
O trabalho trata da criação de um programa elaborado em ambiente MATLAB que calcula os limites térmicos de projeto de um típico reator a água pressurizada (PWR), que é a temperatura central da pastilha combustível e a taxa de ebulição nucleada (DNBR). Outras distribuições de temperatura e grandezas hidrodinâmicas do líquido refrigerante, como a entalpia e a queda de pressão também são calculadas. O código possui peculiaridades, como o fato de permitir cálculos com combustíveis de UO2 puro e proporções do óxido misto de Urânio/Tório - MOX (U,Th). Estas, além da sua interface amigável com o usuário provam que o código pode ser utilizado em trabalhos de pesquisa , bem como em disciplinas de graduação e pós graduação voltadas ao estudo de termo-hidráulica de reatores nucleares em cursos de graduação e pós graduação de engenharia (nuclear e/ou da energia) espalhados pelos país, como no caso do curso de graduação de Engenharia de Energia da Universidade Federal do ABC, onde é uma disciplina optativa. Para a validação do código foram utilizados dados do reator AP-1000 da Westinghouse. O programa se apresentou com comportamento físico dentro do esperado para o modelo, gerando resultados confiáveis para eventuais projetos de reatores (validado com dados experimentais e outros programas), bem como propicia a alunos uma experiência diferenciada dentro da aprendizagem dos conceitos empregados na área, uma vez que o programa permite uma análise mais profunda de determinados conceitos na área de termo-hidráulica que dentro da aula expositiva e com exercícios convencionais não poderiam ser explorados.
Thermohydraulic analysis of a fuel element of the AP1000 reactor with the use of mixed oxides of U / Th using the computational fluid dynamic code (CFX)
2019 - CUNHA, CAIO J.C.M.R.; RODRÍGUEZ, DANIEL G.; LIRA, CARLOS A.B.O.; STEFANI, GIOVANNI L.; LIMA, FERNANDO R.A.
The present work carried out a thermohydraulic analysis of a typical fuel assembly of the reactor AP1000 changing the type of fuel, of UO2 conventionally used for a mixture of oxides of (U,Th)O2 realizing some simplifications in the original design, with the objective to develop of an initial methodology capable of predicting the thermohydraulic behavior of the reactor within the limits established by the manufacturer. An expression for the power density was determined using a coupled neutronic thermohydraulic calculation; once the final expression for power density was determined, the axial and radial temperature profiles in the assembly, as well as the pressure drop and the distribution of the coolant density, were evaluated. Due to the increase in research done on thorium, such as the work of [1], [2], [3], [4] and [5], as well as the mass diffusion of the AP1000, as is the case with [6] and [7]. The present study developed a simplified model, where burnable poisons and spacer grids were not considered, however, it is a consistent model, but with the insertion of these, a more accurate representation of the reactor is expected, providing operational transient analyzes. This tends to strengthen the lines of research that have been carrying out work on the AP1000, as well as in the general sphere of nuclear power plants.
Analysis of a pressurized power reactor using thorium mixed fuel under regular operation
2019 - GOMES, DANIEL de S.; STEFANI, GIOVANNI L. de; OLIVEIRA, FABIO B.V. de
This work discusses a parametric study applied to nuclear power generation based on a mixed fuel formed by the composition of thorium-uranium oxide (Th-U)O2. Also, approached in this study the physical neutrons models of a fuel system composed of ThO2 75 wt% and UO2 25 wt%, with 19.5% enrichment of U-235. The thermodynamic features of the thorium-uranium fuel system compared with the properties of uranium dioxide. Thorium-based fuel operating extended fuel cycles reach of over 80 GWd/MTU in a pressurized water reactor (PWR). Homogenous distribution of thorium-based fuel, used on the reactor core, could reduce Pu-239, once U-233 production capacity dependent on Th-232 replacing U-238 in the fuel matrix. The mixed oxide fuel has a lower buildup of Pu-239, causing the linear heat rate distribution slope to flatten and lowering fuel porosity. The release of gaseous fission products models for (Th-U)O2 could have different diffusion coefficients when compared to uranium oxide models. Besides, resulting in lower thermal gradients than UO2 and a reduction in fuel swelling. This parametric study reviews the aspects of radioactive decay chains of uranium and thorium. It founded the simulation using approved nuclear codes, such as SERPENT for neutron physics calculations and the FRAPCON code, which defines the licensing process. The results show that thoria based fuel has a higher performance than UO2 fuel in regular operation and can improve safety margins.
Behavior of thorium plutonium fuel on light water reactors
2019 - GOMES, DANIEL S.; SILVA, ANTONIO T. e; OLIVEIRA, FABIO B.V. de; LARANJO, GIOVANNI S.
Designs using thorium-based fuel are preferred when used in compliance with sustainable energy programs, which should preserve uranium deposits and avoid the buildup of transuranic waste products. This study evaluates a method of converting uranium dioxide (UO2) to thorium-based fuel, with a focus on Th-Pu mixed oxide (Th-MOX). Applications of Th-MOX for light water reactors are possible due to inherent benefits over commercial fuels in terms of neutronic properties. The fuel proposed, (Th-Pu)O2, can be helpful because it would consume a significant fraction of existing plutonium. Aside from the reactor core, the proposed fuel could be useful in existing technology, such as in a pressurized water reactor (PWR). However, licensing codes cannot support Th-MOX fuel without implementing adaptations capable of simulating fuel behavior using the FRAPCON code. The (Th-Pu)O2 fuel should show a plutonium content that produces the same total energy release per fuel rod when using UO2 fuel. Thorium is a fertile material and demands a slightly higher plutonium content when used in Th-MOX. Mixed ceramic oxides show thermodynamic responses that depend on the comprising chemical fractions, and there is little information in databases on irradiation effects. The neutronic analysis is carried out using the SERPENT code to quantify transuranic production and compare this production with the original UO2 fuel assembly. Parameters such as delayed neutron fraction and temperature reactivity coefficient are also determined. Through these analytical methods, the viability and sustainability of the proposed new fuel assembly can be demonstrated in a closed fuel cycle.
Optimization on the core of IEA-R1 research reactor for enhance the radioisotopes production
2019 - STEFANI, GIOVANNI L. de; GENEZINI, FREDERICO A.; MOREIRA, JOÃO M. de L.; SANTOS, THIAGO A. dos
In this work a parametric study was carried out to increase the production of radioisotopes in the IEA-R1 reactor. One of the variables directly proportional to isotope production is the magnitude of the neutron flux in which some material is exposed, so the main objective of this work was to increase neutron flux, especially in the center of the reactor in the beryllium element irradiator (EIBe), within the operational and safety limits of the reactor. The study is initiated by defining a default configuration, in which core of the IEA-R1 reactor is modeled with all fresh fuel assemblies to ensure the reduction of variables that affect the data analysis, then para metric studies were performed evaluating, by comparative analysis of the behavior of the relation of neutron flux versus the fuel for the standard configuration. Therefore, another configuration was tested: the changes in the core of graphite reflecting elements for beryllium, as well as, the result due to reactor core compaction. Parameters such as the fraction of delayed neutrons (Beff) and temperature reactivity coefficient are analyzed to ensure that the configuration has the minimum safety requirements for the reactor safe operation. The results of the study demonstrate a large increase in neutron flux magnitude and in particular in the center of the nucleus in the beryllium irradiating element.
Detailed neutronic calculations of the AP1000 reactor core with the Serpent code
2019 - STEFANI, GIOVANNI L. de; MOREIRA, JOAO M.L.; MAIORINO, JOSE R.; ROSSI, PEDRO C.R.
In this work we present some validation results for reactor core modeling with the Serpent code performed for the first cycle of the AP1000 reactor. The comparison with reported values of the assembly k∞ for cold zero-power condition showed a discrepancy of 0.29%. The kef for full-core static and burnup calculations of the very heterogeneous AP1000 reactor core also presented good agreement with reported values. The kef for states with uniform fuel and moderator temperature distributions showed discrepancies below 0.91%. The boron worth curve obtained from burnup calculations with the Serpent code model results reproduced very well literature results despite using uniform temperature distributions in the modeling. In addition we discuss shadowing effects among burnable absorber rods (IFBA and Pyrex) and control rods which are, together with soluble boron, the control means throughout the first cycle. For instance, the presence of 9 Pyrex rods in an assembly decreased the average reactivity worth of one IFBA rod from 147 pcm to 33 pcm; and the presence of 28 IFBA rods in an assembly decreased the average reactivity worth of one Pyrex rod from 631 pcm to 277 pcm. The reactivity worth of a black control rod reduces about 20% when 28 IFBA rods are inserted in the fuel assembly.
The utilization of thorium in Small Modular Reactors – Part I
2018 - AKBARI-JEYHOUNI, REZA; OCHBELAGH, DARIUSH R.; MAIORINO, JOSE R.; DAURIA, FRANCESCO; STEFANI, GIOVANNI L. de
This work presents a neutronic assessment to convert a Small Modular Reactor (SMR) with uranium core to the thorium mixed oxide core with minimum possible changes in the geometry and main parameters of SMR core. This option is due to most of SMR are designed to be strongly poisoned in the beginning of cycle and to have a long cycle. Thorium can be used as an absorber in the beginning of the cycle and also be used as a fertile material during the cycle, it seems to be a good option to use (Th/U)O2 as SMR’s fuel. The main neutronic objectives of this study is achieving longer cycle length for SMR by using the minimum possible amount of burnable poison and soluble boron in comparison with reference core. The Korean SMART reactor as a certified design SMR has been chosen as the reference core. The calculations have been performed by MCNP code for homogeneous and heterogeneous seed and blanket concept fuel assemblies. The results obtained show that the heterogeneous fuel assembly is the one which gives longer cycle length and used lower amount of burnable poison and soluble boron, and also consumes almost the same amount of 235U.
Consolidation of the new nuclear calculation methodology of the IEA-R1 reactor
2017 - STEFANI, GIOVANNI L. de; CONTI, THADEU das N.
The IEA-R1 neutron and thermo-hydraulic calculation methodology is composed of 5 computational codes from the area of reactor physics, which have a symbiotic dependence on each other. Since the outputs of each code will be used to generate the input of the next code. The programs involved in this methodology are LEOPARD, HAMMER-TECHNION, TWODB, CITATION and COBRA. Each of these codes is responsible for a specific type of calculation. In a first two-year study, between the years 2008 and 2010, these IEA-R1 nuclear reactor codes were integrated into a single management code. This management code had as main objective the reduction of the time spent by the calculation team of the reactor and to prevent against errors in the manipulation of data and data output. In this study the calculation time was reduced by 99%. The present article presents the closing of this work, being a document with the consolidation of 7 years of the use of the new calculation methodology implemented in the IEA-R1 reactor, demonstrating its efficiency and reliability, besides the proper registration of this project that had great importance Within the IPEN research reactor.